AE-104

o

w<

Problems in Pressure Vessel Design

and Manufacture

0. Hellström Ragnar Nilson

AKTIEBOLAGET ATOMENERGISTOCKHOLM SWEDEN 1963

AE-104

PROBLEMS IN PRESSURE VESSEL'DESIGN AND MANUFACTURE,

Oa Hellström Ragnar Nilson

UDDEHOLMS AB, DEGERFORS JÄRNVERK AB ATOMENERGIDegerfors Stockholm

Abstract

The general desire by the power reactor process makers to in-crease power rating and their efforts to involve more advanced thermalbehaviour and fuel handling facilities within the reactor vessels areaccompanied by an increase in both pressure vessel dimensions andvarious difficulties in giving practical solutions of design materialsand fabrication problemsa In any section of this report it is emphasizedthat difficulties and problems already met with will meet again in thefuture vessels but then in modified forms and in many cases more per-tinent than before» As for the increase in geometrical size it can bepostulated that with use of better materials and adjusted fabricationmethods the size problems can be taken proper care of« It seems likelythat vessels of sufficient large diameter and height for the largest poweroutput, which is judged as interesting in the next ten year period, canbe built without developing totally new site fabrication technique» It is,however, supposed that such a fabrication technique will be feasiblethough at higher specific costs for the same quality requirements asobtained in shop fabrication»

By the postulated use of more efficient vessel material withprincipally the same good features of easy fabrication in'»differentstages such as preparation, welding, heat treatment etc as ordinaryor slightly modified carbon steels the increase in wall thickness mightbe kept low. There exists, however, a development work to be donefor low-alloy steels to prove their justified use in large reactor pressurevessels,1

Paper presented to the EAES Symposium "KEY PROBLEMS OF NUCLEARPOWER STATIONS LIKELY TO BE BUILT DURING THE NEXT TEN YEARS"at Cannes, France, 8 -12 October 1962.

Printed and distributed in May 1963'.

CONTENTS

Page

Introduction 3

Survey of Swedish Reactor Vessels 6

1. Åge sta PHWR Vessel 6

2. Marviken BHWR Vessel 9

3. Project study PHWR-400 Vessel 11

4. Project study Bashful-500 Vessel 12

Vessel steels and strength evaluation 14

Fabrication in shop or at site 1 8

Special shop requirements due to siEe, weight and design

of the vessel 19

Shop fabrication and inspection 20

Summary 24

Figures 1-27 I - XVIH

REFERENCE REMARK

This paper is of course only one contribution among many to a morecomplete picture of the Swedish development work for nuclear powerstations. At the Cannes symposium other important developmentproblems are stated in the following paper: "Major Problems andDevelopment Tests for Heavy Water Pressure Vessel Reactors"by P H Margen.

- 3 -

Introduction

The nuclear power reactor development work in Sweden has since

1958 been lined up with practically total concentration upon heavy water

reactors of the pressure vessel design type. The first demonstration

of this reactor type in Sweden, the Åge sta PHWR reactor is now in the

later stages of construction and will finally be put in order for critica-

lity and the following research program of first core physics and thermal

behaviour and further stepwise increasing power production tests in the

beginning of 1963, This reactor has been described at different occasions

during the last years at conferences and in technical magazines and in

this report design features by its vessel are given mainly

in order to create a reference point in the discussion of design and ma-

nufacture problems by large reactor pressure vessels. Problems of

various kinds which has been treated and brought to practical solutions,

in the different stages of the work upon the Ågesta vessel are doubtless

forming a fertile ground in the work with the following pressure vessels

which since some time are brought to more or less detailed pre-project

studies for our future power reactor program. From the general point

of view the Ågesta reactor - in itself too small and in many features

suspected to be too complicated and technologically too "luxurious" to

be claimed an economical power production unit - might be regarded

as starting point for two attractive outlines of pressure vessel unit

developments for competitive nuclear power, namely

a) the uniform lattice reactor ("homogenized"), PHWR, and

b) the natural circulation direct cycle boiling reactor equipped with

facilities for nuclear superheating, BHWR,

Both these reactors are studied for electricity outputs up to the

order of 400 to 500 MWe and the consequenses in design and manufac-

turing of the adequate pressure vessels are considered. It can be seen

from the following report, that even if the Ågesta vessel must be re-

garded as large and heavy the vessels required for the future reactors

are still larger. With respect to most factors which influence the

_ 4 -

judgment of their feasibility really important extrapolations of the

engineering achievements of today have to be established»

In the question of which reactor pressure vessel is likely to be

built next in Sweden it might be pointed out that a redesign work of

the Marviken Power Plant according to an BHWR lay-out is now going

on for an output size of 200 MWe - when internal nuclear superheating

is included - and that a formal decision on construction can be expected

during the first months of 1963. Planned date for power operations

o£ this unit is 1968. Full scale units - about 500 MWe output - of BHWR

or PHWR might in Sweden be requested for commercial operation in

the beginning of the \c)10'fst

Before going over to the more detailed descriptions of the different

problems it might be proper to lay down some background aspects for

heavy vessels for nuclear reactor purpose. Design, manufacturing,

inspection and use of pressure vessels for different purposes are in

most countries subjected to more or less rigorous regulations and the

application within a new field is consequently among other always re-

garded from the "Pressure Vessel Code" point of view. In 1956 the

regulation aspects in USA were brought to a more common formal basis

by the establishing of the ASME Boiler and Pressure Vessel Code In-

terpretation Case No 1224 (Special Ruling), which then time to another

was completed by definitions and addenda.

In the beginning of 1958 the tentative regulations were brought

together in Case No 1234, which gives the Code aspects upon all pressure

vessels working in the primary reactor circuit or aside from that in such

circuits which still carry reactor coolant containing lethal substances.

Quotation from the interpretation Case No 1234 is made here:

"inquiry: Under what special rules shall a nuclear reactor vessel

or a primary vessel, as defined in Case No 1224, be built in order to

be built in order to be acceptable for Code construction?

- 5 -

Reply: Pending development of more complete rules to cover

nuclear vessels, it is the opinion of the Committee that a reactor

vessel or a primary vessel shall meet the requirements of this Case

in order to meet the intent of the Code and to be stamped in accordance

•with Case No 1224. Where differences exist the requirements of this

Case take precedence over the Code rules for the subjects covered.

The requirements of this Case are as follows? (t), (2) -----(6), (7)

and (8)."

The design calculation, general specification of materials, wel-

ding fabrication and inspection rules given in the items (1) to (6) are

quite well defined additions to and restrictions of normal Code regu-

lations and do not cause much trouble in their interpretation and

technical application.

The items (7) and (8) are of quite another character and do in

fact contain general descriptions of what to be taken special care of

when designing advanced vessels, specially nuclear vessels. Quota-

tion follows here;

"(7) The Code rules are intended to provide minimum safety

requirements for new construction, and not to cover deterioration

which may occur in service as a result of corrosion, erosion, ra-

diation effects, instability of the materials or operating conditions

such as transient thermal stress or mechanical shock and vibratory

loading; nevertheless particular consideration shall be given to these

effects with a view of obtaining the desired life of the vessel.

(8) In view of these severe service requirements, particular

consideration shall also be given to materials, construction, and

inspection, including supplementary methods of non-destructive

testing, so that soundness and good practice will result. Due regard

shall be given to such items as smoothness of welds and to location

and detail of structural attachments. "

- 6 -

Since 1958 there has of course been added more detailed pro-

visional rules and recommendations both by the ASME Boiler and

Pressure Vessel Committee and by safeguard authorities and

societies in other countries,for instance by Lloyd's Register of

Shipping, Land Division, in the Provisional Requirements, 1960.

These rules - later on collected for instance within the ASME

Nuclear Code Cases No 1270 N, 1271 N, 1272 N, 1273 N, 1274 N

and their addenda - do still mostly refer to problems of the same

defined character as covered by the items (1) to (6) above and the

items (7) and (8) still are in principle as generally given as in 1958.

It is easy to see that the problems which are not distinctly covered

by Code rules may cause the engineer more trouble than those which

are covered. It is further quite obvious that the questions which are

not covered by Code rules may have a great and perhaps dominant

effect upon fabrication and construction costs.

As the evaluation of all these questions on the basis of experienced

correlations between used design and intended vessel usage and real

vessel behaviour during operation still is poor or lacking, it must be

stated that sufficient background for a verified economically optimized

design work does not yet exist.

Survey of Swedish Reactor Vessels

A schematic study in size growth for these vessels is demonstra-

ted in Fig. 15 on figure page XIII after the text.

The Åge sta reactor vessel is shown in Fig, 1, 2, 3 and 4 - figure

pages I - IV after the text - and the principle data are to be found in

table 1 at the end of this survey chapter on page f 3. Specially interes-

ting parts of the vessel are:

- 7 -

Bottom dome is of ellipsoid! c al shape with the main nozzles for

inlet and outlet of the coolant. Originally this dome was assumed to be

made with a spherical crown radius for the central portion and with a

torus radius for the outer knuckle portion. From extensive strain

measuring on a scale model, however, it was gathered that the com-

bination of large nozzles and their full reinforcement would give

stresses of somewhat irritating magnitude for a nuclear vessel. As a

result the shape was changed to ellipsoidical which was judged to give

diminished stresses particularly around the nozzles. The theoretical

examination was confirmed by the strain measurement results obtained

for the full scale vessel in connection with the final pressure test.

The flat top cover is equipped with light water circulation for

balanced heating and cooling guided by and controlled against the reac-

tor coolant temperature. The height of top cover is dictated not only

by the need to keep weld seam stresses low but also by the requirement

that radiation dose should be kept at low rates for the upper closing

plate and the tightening flange so that the top plate could be accessible

whenever the reactor is shut down.

The top cover is penetrated by standpipes for fuelling operation

and application of control rod drives. As this reactor has intermittent

fuelling operation at rather short intervals the standpipe closures will

be handled quite many times (roughly 200) during the desired total opera-

tional lifetime. If,these closures are maintainable at non-leakage con-

ditions will be an interesting experience.

The special flange design - the type is shown in Fig. 8 - which

permits reactor starting-up and shut-down operations performed within

10 hours each without producing any overstresses in neither the

vessel nor the top cover, The flange in itself is tightened by a primary

lensformed silver gasket and a secondary seal-welded torus.

All the internal vessel equipment - shielding komponents and

water distribution box etc. - is adapted for remote handling by special

- 8 -

decomposition and lifting gages and the vessel itself can be remote

control lifted out of the concrete biological shielding.

A delicate problem was formed by the demand for very narrow

tolerances for vessel roundness and axial alignment.

All heavy water wet surfaces are claimed to be stainless steel

clad and surface finish in order to prevent contamination was demanded

to be high quality, microdepth of profile in the order of 0. 016 mm. The

stainless cladding was permitted to be supplied as roll-clad or as

deposit weld clad,

A special feature which might be regarded as quite conservative

is the arrangement of neutron and thermal shielding. The radial shiel-

ding is made from stainless steel cylindrical shells concentrically

placed with a very thin annular ring space between and are together

forming 1 50 mm steel between the outer radius of the core reflector

and the vessel wall. Thus the integrated irradiation dose on the wall

will be quite modest even for a core of higher thermal power than the

actual one. Furthermore special channels for boxes containing test

pieces of vessel steel are arranged in the outer portion of the radial

stainless steel shielding. This boxes are remotely handled by a special

machine for removal and transport to test facilities, where control

results of irradiation embrittlement of the vessel steel can be obtained

from time to time.

In order to verify intended design decisions different component

investigations were carried out. One of these investigations is represen-

ted here by Fig. 10, showing scale model of the vessel. That model

was investigated for stress concentrations in the different parts of

tank and lid and subjected to pressure cycling in order to verify the

leak tightness of the primary silver gasket arrangement. Bottom dome

shape modification was initiated by the test results. Another study is

shown in Fig. 11, which illustrates a detailed study of a flat lid portion

- 9 -

including the part of the standpipe which penetrates the lid. Investiga-

tion of stresses and deformations behaviour under thermal cycling and

examination of welding procedure were the preset aims of this work,

which gave some valuable hints to detail modifications.

2. _ Marviken BHWRJVessel

The Marviken reactor is just now - as mentioned in the intro-

duction of this report - in a redesign period. The reactor vessel is

shown in Fig. 5, 6, 7 and 8 - figure pages V - VIII after the text -

and the principal data are listed in table 1 on page T 3, All features

given in this report belong to the redesign version although some

of them, e.g. the internal fuel element manipulator, which gives

the vessel height, can be found in an earlier PHWR version for Mar-

viken.

While the Ågesta vessel has a lot of standpipes for refuelling

the Marviken vessel is equipped with only one loading channel pene-

tration. As the control rod hydraulic drives are arranged internally

at the top of core shroud, penetrations for these drive units are

eliminated from the vessel itself. The cylindrical vessel shell is

penetrated by feed water piping for control rod drives, medium sized

steam piping for pressure relief valves, small size piping for emer-

gency cooling of boiler fuel elements and fuel canning leak detection

and medium sized piping for emergency cooling of superheater fuel

elements. Those penetrations can be assembled within a rather loca-

lized cylinder shell ring portion which has to be reinforced by increase

of wall thickness. Outside of that region the cylinder shell is quite

clean and straight.

The severe penetrations problems are concentrated to the bottom

dome where 30 superheater nozzles and some other single position

penetrations are arranged. The bottom dome design and manufacture

will undöubtly form a work of highest,qualification and careful studies

of what should be the detailed technical decisions are underway. Among

- 10 -

those studies special attention is paid to the superheater nozzle itself,

illustrated in Fig. 7, Calculation - eventually followed by model tests -

upon thermal charge and temperature distribution of the nozzle are

carried out intensely and the consequences for the steam flow correctly

balanced isolation from the nozzle body are estimated. The design

involves also some rather intricate problems of compatibility between

different structural materials.

The main design problems with this large vessel can - aside

from the special bottom dome problems - be listed as follows

a) demonstration of the use of low-alloy vessel steel in order to reduce

vessel weight and to give medium wall thickness for high quality

welding and inspection,

b) realistic efforts to get away from specifications of costly narrow

tolerances in shape and measures,

c) liberalizing of the restrictions upon cladding materials and surface

finish quality,

d) the application of a suiting flange design in view of the fact that

the main tightening will be opened only for major rearrangements

of the internal reactor equipment, which cannot be carried out by

the system of internal manipulator and outer facilities for fuel

element and control rod handling. In this question it might be noticed,

first, that the Åge sta flange type according to Fig, 8 is not as fully

justified by this vessel behaviour as by Åge sta with its flat lid design,

second that the Agesta type seems to be somewhat unfavourable in

dimensions and weight, especially for the bolts, at this vessel size

and higher pressure, and third that the care against thermal transient

loads and moments in the flange region of the vessel would cause a

need for development work upon principally new flange types. Such

types are studied since some time and good prospects of a flange

design, in which the influence of excentric loading and thermal

forces is eliminated, seem to exist.

- 11 -

In this reactor vessel the neutron and thermal shielding is acquired

by a somewhat different arrangement than by the Ågesta vessel. Theore-

tical and experimental research has given verifying contribution to this

arrangement, which will use a comparatively thin radial stainless steel

shield combined with a thicker layer of heavy water between the steel

shield and the vessel wall. The efficiency of this method seams to be

superior to that of using thick stainless steel shielding with exception

for the fact that an increase in heavy water inventory could result.

The width of the annular ring space for heavy water outside of the steel

shielding is, however, in this case also needed for the thermal process

water flow. It can be foreseen that the arrangement of test specimen

channels for time to time watching of the irradiation damage on vessel

wall steel would still be kept as an operation condition for this vessel

as it is for the Ågesta vessel.

;PHWR:400Ve s sel

A principal drawing of this pressure vessel is shown in Fig, 9 -

figure page IX after the text - and the principle data are listed in table 1

on page 13 . Special features for this vessel are

a) top dome lid is to be removed at each fuel handling occasion as the reactor

is projected to have a fuel handling procedure similar to that of light

water reactors,

b) hemispherical bottom dome design requires some considerations with

respect to the main inlet and outlet nozzles which are positioned up

in the cylinder flange region of the dome. The control rod drive pene-

trations in the central zone of the hemisphere would not give major

trouble.

As these technical applications have been verified in rather many

cases before this study it seems correct to postulate that most of the

considerations will be i*eiterated verification of the soundness and good

sense in design modifications initiated by different manufacturing methods.

- 12 -

In addition some development work might be included upon more exact

calculation methods for reinforcements in transition regions disturbed

by bending moments.

Ye ssel

The reactor vessel of this study appears in this report mostly in

table 1 on page 13, where the principal data are given for comparison

with Marviken BHWR,

Table 2 on page T7 shows some speculations around needed vessel

wall thicknesses by the use of different vessel steels.

- 13 -

Table 1. Survey of Reactor Data

Core diameter mCore height m

Radial reflector mTop " mBottom " mCore thermal output MW

Design pressure bar" temperature C

Saturated steam temp. C

Superheated steam " C

Shielding radial thickness inStainless steel mm

Heavy water mm

Irradiation flux atvessel wall n/cm sek

(E - 1 MeV)Vessel inner diameter mCylindrical shell

thickness in vessel•steel mm

Vessel steel '

Vessel height mControl rod positionsFuel element handling

penetrationsSuperheater penetrations in

bottom dome

1) Steels according to Table 2,2) Estimated values3) Not penetrating the vessel

Åge staPHWR

3, 613.04

0 . 3

0 . 3

0.3rf» ,

/ 65\(125)

40

251

150

21

1.2* 109

4.555

65

carbon-manganese

9.5

27 + 2

37

page 17.

MarvikenBHWR

4,905.02

0 . 3

0 . 3

0 . 3

570

57.5

272

263

500

50

110

92 )3-4 • 1<K

5.22

61

low- alloy

22. 1303>

1

30

ProjectStudyPHWR-400

4, 66

5.50

0 . 3

0 . 3

0 . 3

1200

80

293

90

130

92)4. 2 • 107

5. 10

90

low-alloy

14. 0

12

Ttop lidY removal

ProjectStudyBashful-500

4.83

4.83

0.330.330.331360

77

291

283

500

(-70)

(-45)

92)-4-6* 107

5.72

98

low-alloy

24.0373)

1

40

- 14

Vessel steels and strength evaluation

In Fig. 12 the specification for the Ågesta vessel steel is given and

in Fig. 13 the impact qualities of the virgin material before fabrication

are demonstrated. From most aspects this steel is quite wellknown

beeing a fine grain treatment modification of a Swedish standardized

pressure vessel steel. The requirement of extraordinary low cobolt

content for nuclear use is representative for the period when the Ågesta

specification was decided. From table 2 and Fig, 14, however, it can

be read that the continued use of a vessel steel with such modest strength

values would cause rattier high wall thickness and as a result of that

quite impossible vessel component weight especially by the larger

vessels. The welding procedures and effective inspection possibilities

are also seriously negatively affected by increasing wall thickness. The

evaluation of thermal stresses resulting from blocked transient thermal

strains is also considered to give strong arguments against thick ma-

terials. It is then an automatic consequence that steels with higher

strength values and still preserved good behaviour in fabrication and

use must be given keen attention. The strength calculation values,

38 - 41 kp/mm at actual design temperatures, given in table 2, are

representative for low-alloy vessel steels obtainable from Swedish

steel-makers today. For the Marviken BHWR reactor vessel the design

and manufacturing pretest work with the Fortiweld Normal steel is

now going on. This is a boron micro-alloyed fine grain treated carbon

steel and gives in normalized condition a yield point value of minimum

45 kp/mm at room temperature and is since some years used in

comparatively large conventional pressure vessels. As for Charpy

impact values this steel is specified for guarantee of 3, 5 kpm/cm

(25 ft. lbs) at 0 C, For ordinary carbon vessel steels the specified

Charpy impact value according to the Swedish Pressure Vessel Code

is 2,6 kpm/cm2 (15 ft.lbs) at -20 °C.

As can be computed from estimated neutron flux data in table 1

the integrated irradiation damage dose (E > 1 MeV) upon the vessel

wall for the Marviken BHWR reactor would be in the order of 3 * 10 n/em

- 15 -

at 30 years total life time and utilisation rate of 80 %. According to

preliminary results from irradiation tests upon this steel this inte-

grated dose might give a fairly large contribution to the rise of brittle

fracture propagation transition temperature. It is today a bit early

to state wether this would cause special requirements as to the

operation conditions for the reactor. If, however, the safeguard

evaluation of the embrittling behaviour would tend to give severe

restrictions in the operational schedule it seems likely that a

medium-strength vessel steel would be chosen this time. Basical

experiments and evaluation work have to be done in order to establish

more explicit use of the scattered and un-correlated knowledge of

irradiated steel behaviour.

The calculation and evaluation of strength conditions for the

ductile behaviour of reactor vessel steels are in some respects

carried out with more care than what is prescribed by pressure vessel

codes applied upon ordinary vessel with dangerous contents. With

respect to undisturbed membrane stresses and mean stress values

the considerations have given the result that no argument exists for

a more conservative attitude. The unofficial provisional requirements

used as additions within this work are specifically valid for thermal

transient stresses and bending stresses. The following list of factors

which can form different patterns for the evaluation of safety factors

gives the situations

Type of loading,

Uncertainly in load evaluation,

Inherent stress conditions,

Real correlation stress verses strain,

Multiaxial stress-strain correlations (plastic yield behaviour criteria),

Deformation hardening and Bauschinger effect,

Relaxation (creep at low temperatures),

Low cycle fatigue,

Corrosion,

Normally existing material defects,

Faultive material,

Influence of fabrication tolerances,

Compatibility between different materials,

Degree of statical indeterminate condition,

Uncertainly in computing methods.

The value of such a xather comprehensive evaluation model de-

pends entirely upon, first real knowledge of reactor operating con-

ditions and second,upon well correlated practical and experimental

experience from vessel behaviour as well in design details as in

material characteristics.

- 17 -

1.

2 .

3 .

4 .

5.

6.

Design pressure

Design temperatureInner diameterVessel steel type

Cylindrical wallthickness -.'(exclusive claddingmaterial)

Bottom dome wallthickness

ALTERNATIVE

5.

5.

t)2)

3)

4)

5)

6)

1 Cylindrical wallthickness for steelaccording to note 1)

2 Cylindrical wallthickness for low-alloy steel accordingto note 4)

Table 2.

bar

°Cm

mm

mm

mm

mm

Reactor

ÅgestaPHWR

40

250

4.555

Vessel Dala

carbon- ,\manganese '

65

65

65

Stress calculation value at actual designU It II

n it it

M If It

Hemispherical shapeAllowable hoop stress

ti i

t) !

t! I

t n

i n

PHWR--400

80

293

5. 10low- .alloy '

90

905>

165

75

temperaturei i

n

M

calculation according to theSwedish Pressure Vessel Code, 1959.

Marviken--BHWR

57.5

272

5.22

low- 3 ,alloy >

61

95

114

52

24 kp/j

38 "41 »

46 "

Bashful--500

77

291

5.72lOW- oy

alloy L)

98

165

180

80

2tim

- 18 -

Fabrication in shop or at site

The size development of heavy water reactor pressure vessels

in Sweden during the next ten years has already been described, Fig. 15,

The rapid increase of the over all dimensions and weights needs a care-

ful investigation of the suitableness of fabricating the vessel in shop, at

site or a combination of the two. A survey of the requirements for a

heavy water reactor pressure vessel as to weld procedure, machiningj

tolerances, final assembly, testing and adjusting gives a distinct ad-

vantage for shop fabrication. Further it is good reason for a statement

that a complete site erection from pressed plates similar to the fabri-

cation of gas cooled reactors in England is more or less impossible.

These facts give rise to the conclusion that a combined fabrication in

shop and at site presupposes that the site must be equipped with a

fabrication shop for putting together largest possible prefabricated

components. Such a shop at site must have facilities for welding,

machining, lifting, stress relieving etc. A good example of this

combined fabrication is the twelve 350 tons heat exchangers for

Trawsfynydd in North Wales.

A complete shop fabrication gives immediately rise to the problem

of transporting the vessel from shop to site. The original intention as

to the Agesta pressure vessel was to transport it a short distance by

rail to the coast and then ship it to site, A combination of delivery date

and the special ice conditions in Sweden did change that scheme to a

complete railway transport. The departure from Degerfors of the

pressure vessel, the top cover and the outer thermal shield is shown

in Fig. 16. This transport item with a diameter of more than 5 meter

and a weight of approximately 200 tons gave one of the most complicated

transportations ever made in Scandinavia and is impossible to repeat

for the much bigger vessel to Marviken, where the original plan for

Ågesta has to be used with the modification that the Marviken vessel

must be floated to site. It is also possible to predict that it is very

likely that such a transportation scheme even can be used for a

600 MW BHWR pressure vessel, which conclusion is based of to-day's

- 19 -

knowledge of design and steel properties. Consequently, for the next

TO years reactor construction in Sweden it seems possible to stick to

shop fabrication without serious extrapolation of to-day's knowledge,

provided that the site is situated on or close to the coast and that the

vessel can safely be taken out of the water and moved to site.

Special shop requirements due to size, weight and design of the vessel

The fact that shop fabrication is a distinct advantage combined

with the above conclusions regarding transport increases the interest

to point out the special shop requirements to make a proper job. The

Marviken pressure vessel with an assembly height of 22. 1 m, a maxi-

mum diameter of almost 6 m and with requirement for a maximum lift

of almost 200 tons gives a very clear answer as to the shop dimensions,

overhead crane capacity and facilities for forming, welding, machining

and stress relieving.

There is no doubt about that there is an increased demand for

heavier pressure vessels from the chemical process industry, oil and

petrochemical industry etc., but the above mentioned example of require-

ments is - generally speaking - ja front of the normal development owing

to the fact that such sizes usually can be prefabricated in shop and put

together at site.

One of the main differences between a heavy water reactor pressure

vessel and an ordinary vessel of the same size is the tolerances. For

the Agesta reactor vessel the shell radius tolerance requirements was

- 2 , 5 mm on a radius of 2, 3 m e. g, 0, 2 %, which is 5 times closer

than the code rules in Sweden and USA prescribe for an ordinary vessel.

It is obvious that these close tolerances require special tools for for-

ming, special studies and equipment for welding and special purpose

machining facilities.

Due to the fact that the design and the method of fabrication do not

permit an adjustment of a mistake require special equipment and trained

- 20 -

engineering capacity in order to carry out full scale protesting. A

good example is the fabrication of the flat top cover for Åge sta, which

is a complicated honeycomb structure of 50 mm steel plate. One stage

of the fabrication is shown in Fig. 17. When the lower closing plate

was finally combined with the grid it was neces.sary to carry, out the

welding with aid of a mirror , which is shown in Fig, 18, A repair

of these welds was impossible. Of that reason a thorough study was

made as to the dimensional changes and weld quality on full scale test

specimens before the fabrication started. Some more examples are

given below.

Shop fabrication and inspection

. A statement that a thourough knowledge of the material properties

and behaviour during shop fabrication is true in all pressure vessel

production. It is no overstatement to say that this knowledge is more

important in a fabrication of a heavy water reactor pressure vessel

than it normally would be. Here it is only possible to give some examples

such as hot deforming properties and spring back by bending and pressing;

dimensional changes by cooling from hot forming, welding and stress

relieving; changes of mechanical properties in weld metal, plates and

forgings by heat treatment. Thesa examples point out the necessity for

a close contact between a well equipped and highly trained laboratory

and the shop fabrication in order to be sure that the final product is

in full accordance with the specification as to design and mechanical

properties.

It can be worth mentioning that in order to be able to judge

possible changes in dimensions and material properties during the

fabrication of the Agesta pressure vessel a series of full scale tests

were made. These included for example:

1) Hot pressing and heat treatment of 85 mm clad steel plate for the

outer shell of the top cover,

2) Cold pressing and stress relieving of 70 mm plate for the shell,

3) Hot pressing, heat treatment etc. for a 100 mm steel plate for the

bottom closure.

- 21 -

The above mentioned tests verified, the methods to keep the ma-

terials properties according to the specification. On the contrary the

welding tests showed that some alterations from normal procedure

had to be made and that new welding methods sometimes had to be

developed.

Some fabricating problems during a shop fabrication of a heavy-

water reactor pressure vessel will be illustrated in the following

by giving a short description of some production stages of the Ågesta

vessel.

The tank bottom started with welding of pressed sections, which

is shown in Fig. 19. This spherical shaped segment was hot flanged

in a gap press, see Fig, 20, This flanging operation had at the same

time to meet the above mentioned close diameter tolerance for the

vessel and the accurate knuckle radii for the ellipsoidical shape of the

bottom.

The main welding me thods used were manual arc welding and

submerged arc welding. Preheating had to be used all the time except

for the stainless steel welding. The tank bottom being welded on to

the shell from the inside of the vessel by using submerged arc welding

is shown in Fig, 21, where the eight main nozzles for inlet and outlet

of the coolant and the central support and locking for the outer thermal

shield is visible,

A very complicate internal reactor component from the welding

and machining point of view was the water distribution box, which is

shown in Fig. 22,

The necessity for having facilities for stress relieving is illustra-

ted in Fig* 23, where the pressure vessel just is taken out of the furnace.

The top cover is penetrated by a large number of tubes for fuelling

operation and application of control rod drives. All these tubes had to

be welded in and during these operations the top cover was constantly

preheated for about six weeks, A view of the top cover after the tube

welding is shown in Fig, 24, The charging tubes and all small tubes for

control and detecting applications were welded at site. The top cover

after completing of this job is shown in Fig, 25, where the cover is just

going to be placed on the tank portion of the vessel.

The inspection program was very extensive and covered at first

the steel production, plate rollings casting, forging, tubing etc. before

the vessel fabrication started. In this fabrication the inspection covered

weld operator tests, weld procedure tests, X-ray and ultrasonic in-

vestigations and testing of steel coupons for control of the mechanical

properties all the way through the fabrication.

It may be of some interest to mention a special investigation,

which was made on welds representing the submerged arc type in the

tank bottoms Fig» 27, and in the shell of the vessel, Fig, 26,

Fig. 27 shows some results from tensile and shear tests on ma-

terial from a weld, representing the submerged arc welds in the bottom

head. After welding the material has been normalized and stress re-

lieved for a considerable time.

Both tensile and shear tests show that there is no significant

difference between surface and centre of the plate material but that

there are differences in the weld metal. The centre part of the weld

has a somewhat higher analysis according to a greater proportion of

base metal being melted and mixed with the weld metal. Thie gives

the centre part of the weld about the same tensile and shear test

values as the base metal. In the other parts of the weld these values

are lower according to a weaker chemical analysis, but the tensile

test results still meet the UTS-specification of the plate material»

- 23 -

Fig. 26 shows some results from a similar weld that has been

stress relieved only. Here the normal harder peaks in the heat affec-

ted zones can be observed and the weld metal shows higher values

than in the previous case.

Interesting test material was received when eight discs were cut

out in the tank bottom for the penetrations for the main nozzles for the

coolant. This material had under identical circumstances passed the

same operations as the tank bottom followed by the stress relieving

treatment of the vessel» A very thorough investigation was carried out

as to microstructure and mechanical properties of the base metal and

the welded joints. The results confirmed those obtained by the pretests

and can be summarized in the following way,

a) The yield strength of the pressure vessel steel decreases slightly

after hot pressing,renormalizing and prolonged stress relieving

but lies still within the specification.

b) The impact-transition curve for the pressure vessel steel shifts

to a slightly higher temperature by especially prolonged stress

relieving.

c) Tensile specimens taken crosswise to the manual arc welded joint

showed a slight drop in the yield strength compared with the pretests

but the values obtained were still within the specifications. Impact

testing of the weld metal showed that this had kept its good impact

properties.

d) Tensile specimens taken crosswise to the submerged arc welded

joint did not show any difference in yield strength compared with

the pretests". The impact strength of this weld metal was, as the

pretests had shown, inferior to that of the base metal and the

manual arc weld metal, but still within the specification require-

ments.

Surplus material of the above mentioned type and origin is de-

livered to the customer - AB Atomenergi - -who has prepared tensile

and impact tests specimens, which will be put into the eleven tests

- 24 -

specimen channels in the outer thermal shield of the reactor. This ma-

terial will give interesting information of the radiation damage effect

of the pressure vessel steel at different intervals during the life, time

of the reactor.

Summary

The general desire by the power reactor process makers to in-

crease power rating and their efforts to involve more advanced thermal

behaviour and fuel handling facilities within the reactor vessels are

accompanied by an increase in both pressure vessel dimensions and

various difficulties in giving practical solutions of design, mate rials

and fabrication problems. In any section of this report it is emphasized

that difficulties and problems already met with will meet again in the

future vessels but then in modified forms and in many cases more per-

tinent than before. As for the increase in geometrical size it can be

postulated that with use of better materials and adjusted fabrication

methods the size problems can be taken proper care of. It seems

likely that vessels of sufficient large diameter and height for the

largest power output, which is judged as interesting in the next ten

year period, can be built without developing totally new site fabri-

cation technique. It is, however, supposed that such a fabrication

technique will be feasible though at higher specific costs for the same

quality requirements as obtained in shop fabrication.

By the postulated use of more efficient vessel material with

principally the same good features of easy fabrication in different

stages such as preparation, welding, heat treatment, etc. as ordi-

nary or slightly modified carbon steels,the increase in wall thickness

might be kept low. There exists, however, a development work to be

done for low-alloy steels to prove their justified use in large reactor

pressure vessels. The problem of brittle fracture has haunted many

technicians in various fields since it was first observed upon subjects

of rather rough and undefined design and fabrication standard. The fact

- 25 -

that irradiation gives embrittling effects in vessel steels is in itself

out of doubt. The question of what this rise of ductility transition

temperature would really mean to vessels fabricated at the best

practice of manufacturing is still not answered today. The need for

more accurate definition of what safety evaluation against brittle

fracture should cover is obvious.

Another interesting problem complex is connected to the demand

for higher process temperatures and more sophisticated design in

local parts of the vessel. This will quite rapidly bring the aspects

of creep and thermal fatigue within the complex of questions to be

dealt with.

Aside from these advanced aspects the interest in building reac-

tor vessels at lower cost rates will give a full batch of problems even

if the prestanda of the vessel would not be required to increase physi-

cally» Although there may have been no total failures of reactor vessels

in service until now the serious situation is demonstrated by the fact

that during fabrication, inspection and assembly work threatening de-

fects have been revealed. In some cases the observation was made

as late as when the vessel was already assembled to circuitry piping

in the plant. The resulting time élelay and additional money expenses

have of course a tendency to give rise to unfavourable remarks upon

the present aims of nuclear vessel technology. The true situation of

nuclear vessel work of today might be stated so that all the way through

each problem treatment involves the very best of an ever increasing

quality standard practice completed with real extrapolations of the

probability judgment of yet sparely clarified effects of various physical

influences. As often will be the result from handling technical problems

this way the costs have a vivid tendency to go high. Many of the extra

considerations in design and choice of structure materials, fabrication

methods, quality control and pertinent inspection are no doubt justified

by the absolute need for safety against nuclear hazards during an eco-

nomical vessel life time. It could, however, be suspected that the line

- 26 -

of conservative technical decisions in any problem is a bit unbalanced

and not really giving that superior safety, which is justified to pay for.

Thus selective studies of which factors in pressure vessel work do

really inflict upon vessel safety are needed, so that technical decisions

code regulated or not - leading to specifically high costs without adding

any real or very small favour points in safety and long life vessel use

could be eliminated or adequately ranged.

RN/EL

COOLING SYSTEM FOR CONTROL RODS

BORON INJECTION PIPE

_ \ _ -\ PIPE FOR EMTING REACTOR LID

GUID FUNNEL FOR,

FUEL ELEMENT

REACTOR PRESSURE VESSEL2 FILLER RING

3 FILLER BODY

4 OUTER THERMAL SHIELD

INNER THERMAL SHIELD

6 WATER HEADER

7 LEAD OFF RING

8 LID THERMAL SHIELD9 VESSEL LID

10 FLANGE ASSEMBLY

11 LOCKING RtNG

12 CENTRAL BOLT

13 NUT

15 SEALING RING

16 LOCKING PIN

17 GUIDING F IN

18 GUIDING BLOCK FOR VESSEL

19 GUIDING BLOCK FOR BOTTOM RADIATION SHIELD

20 SUPPORT RING

21 SUPPORT BRACKET

SECTION OF ÅÖESTA REACTOR VESSELAB ATOMENERGI

Pressure Vesselwith filler body

III

Fixed Blockfor insulation

Guiding blockfor Vessel'souter ThermalShield

Filler Body(Outer Piece)

(Central Piece)

Outlet Nozzle

Interspece water Intake nozzleInlet

Modified Torv

Bracket forTorus

SuspensionBracket forbottom flangering

Guidance forVessel inBiologicalShield

ÅGESTA REACTOR VESSELTANK DESIGN

Reactor Lid with

Upper Thermal Shield

o oen

V >-3

3 >S w

°0

H

4 Pipes to pressurize*1 system

Loading stand

pipe with

control rod

Thermocouple

Detection Tube

Fuel

Canning Leak

detection

Outlet header for

Lid cooling system

\

Sealing Tube

Torus

Guide for Lid

thermal shield

Holding Bolt for

upper thermal shield

in Lid

Loading Intake header for

Standpipe Lid cooling system

Inspection Tube

Fuel Canning Leak

Detection Tube

Reactor Vessel Wall

Outer thermal

shield

i i

pSERVICE WITHBOILING WITHSUPERHEATING

SERVICE WITHBOILING WITHOUTSUPERHEATING

vVI.

REACTOR PRESSURE VESSEL

RADIAL THERMAL SHIELD

BOILER CHANNELS 151 OFF

SUPERHEATER CHANNELS 30 OFF

CONTROL RODS 30 OFF

t NEUTRON DETECTORS 9 OFF

/T*\ TRANSPORT CHANNEL FOR\\J CONTROL RODS AND FUEL

ELEMENTS 1 OFF

MARVIKEN BHWR LATTICE

AB ATOMENERGI

VII

\ / / /

SUPERHEATED STEAM

5/5 2218

BEFORE THROTTLING

22Cr/f2ri,'WELDED CLADDING

\ \ \ \ \

\ \ \

COMPOUNDED IN 2 MM LAYERSOR k MM WELDED CLADDING

\ \

\ X \ \ \ \ \ \ \

\ \ \ \ \ \ \ \

SIS2324 Ti'rArtAlt /ricoHtL

5/S 2333

ef S/3S333

MARVIKEN BHIVR SUPERHEATER NOZZLES IN

BOTTOM DOME. PROPOSED TEST MODEL DESIGN

VIII.

3/90

3!

-_•=. = -x - .

FLANGE DESIGN OF ÅGESTA TYPE MODIFIED

FOR MARVIKEN BH.VR

IX

Sfoo

J

Preliminary dimensions

PHWR-400 VESSEL WITH MODERATOR

TANK AND SHROUD HEAD

X .

FIG 10

SGALE 1 : 3 MODELOF ÅGESTA VESSEL

FIG 11 TEST MODEL OF FLATTOP COVERÄGESTA VESSEL

XI

FIG 12

CARBON -MANGANESESTEEL, FOR ÅGESTA .VESSEL SPECIFICA-TION

UODEHOLMS A B

Degerlors Järnverk

Carbon-Manganese - Steel2103-R3

ÄgestaNuclearReactor

Composition

C

Si

Mn

p

SCr

Ni

Cl!

Co

N

»At11

max. 0,16 "

0,15-0,5

max. 1,6 »

II 0,02

II 0,03II 0,1

II 0,1

" 0,1.

« 0,020

II 0,015

7.II

II

„n

n

II

|| (200I I

C max. 0,15-Mn max.

" ° ' M ' n

ppm)

1,7

1.8

Mechanical Properties

Yield Strength mm 30 kp/mm(Lower yield point)

(a 30mm)Ultimate TensileStrength 48-58 kp/mm

Elongationon 200mm mm. 18 *U

Impact Strength .(Charpy-V.LS) min. 2,6 kpm/cm

(15 f t lbs)at-20'C

Heat Treatment Nnrmnhcmrj

The Steel is used as:

@Q£li'rLg_DiQt̂ riöL_Q.i EonsLngs^m

Shell of the vesselGrid in the lidUpper plate in the lid Bottom headSupporting ring in 5hell of the lidlocking devise

Flange of the shellFlange of the lidLoose flanges

Lower plate in the lid Nozzles in thebottom head

Imcact properties of 2103-R3 steel.

Normalised materialirnm/rrn Charpy-V specimens\^JI l l rv l l l

2423

2221

20

191817161514

13

1211

109B7

65

432

1

70mm clad steel plateForging 500»320mm

////

/ /

/

//II

/

/: // /

/ // '• /

/ '• // / /

Longitudinal

Transvers

/ ^/

/

//

/f

Temperature C

FIG 13

IMPACT QUALITIES OF THEÅGESTA VESSEL STEELUNIRRADIATED. (FORIRRADIATION DOSE INTE-GRATED- 2xlO18

TRANSITION TEMPERATUREIS RISED LESS THAN 50° C).

-U0 -120 -100 -80 -60 -40 -20 0 +20 +40

XII.

200 mm . .

150 mm ..

100 mm . .

50 mm

Cylindrical shellthickness

Carbon-mangalese steeltable 2 note l)o =30 kp/mmy. p. Vl

Low-alloy steel

table 2 note 2) and 3)

o =45-50 kp/mmy. p. w

Low-alloy steel table 2 note

o =55 kp/mmy. p. F /

4)

(Design pressure) x (inner diameter)

200 300 400 (bar x m)

tÄGESTA

PHWRMAR VIKEN

BHWRPHWR--400

BASHFUL-500

FIG 14 INFLUENCE OF STEEL QUALITY UPONWALL THICKNESS

5

25 m •

20 a -

15 m -•

10 m

5 m

0 m

J )?f/7'

ÅGESTA PHWR PHKR-400 MARVIKEN BHWR

Scale 1:200FIG 15 SCHEMATIC STUD? IN SIZE Ö

HM

XIV.

FIG 16DEPARTURE FROM SHOP OFMAIN PARTS OF THEÅGESTA VESSEL

FIG 17

ÅGESTA VESSEL.TOP COVER HONEY COMB GRID INWELDING POSITIONER

FIG 18

SPECIAL WELDING METHODUSED IN NARROW BACK-SPACEIN TOP COVER.ÅGESTA VESSEL

: Diagram over Ihe principle used for welding in the lid bythe aid of a mirror

xv.

PIG 19

SEAM WELDING OP BOTTOM DOMEPRESSED SECTIONS.ÅGESTA VESSEL

PIG 20

CYLINDER FLANGE HOTFORMING OP DOME.ÅGESTA VESSEL

FIG 21

SUBMERGED ARC WELDING OF BOTTOMDOME TO THE CYLINDER SHELL

XVI

-«• * • • ' « . »*» ^ j ijji» «

w < \

••<j

FIG 22 TEST ASSEMBLY OF FUEL, ELEMENTFUNNELS INTO WATER DISTRIBUTIONHEADER. ÄGESTA VESSEL

FIG 23

STRESS RELIEVINGFURNACE BEHIND THEÄGESTA VESSEL

XVII

vA.

• \

FIG 24 ÅGESTA SITETOP LID BEFOREWELDING ON THESTAND PIPES

FIG 25

ÅGESTA SITESTAND PIPES AND SMALLSIZE PIPING ASSEMBLED.LID PREPARED FORCLOSING REACTORVESSEL.

XVIII.

Results trom shear tests on mikrospecimensfrom a submerged arc weld in plate of 2103-R3 quality.

PIG 26

MICROTESTING OP WELD ANDTRANSITION ZONE MATERIAL.STRESS RELIEVING ONLY

Shear strength

kp/mm

Heat treatment after welding Stress relieving

FIG 27

MICROTESTING OF WELD ANDTRANSITION ZONE MATERIAL.NORMALIZING AND STRESSRELIEVING

Results from tensile and shear lesls on mikro-SD&cimens from q submerged arc weld in a100mm plate of 21Q3-R3 quality

UTS

Heat treatment after welding Normalizing and stressrelieving

LIST OF PUBLISHED AE-REPORTS

1—29. (See the back cover of earlier reports.)30. Metallographic study of the isothermal transformation of beta phase in

zircaloy-2. By G. Östberg. 1960. 47 p. Sw. cr. 6:—.

31. Calculation of the reactivity equivalence of control rods in the secondcharge of HBWR. By P. We'ssglas. 1961. 21 p. Sw. cr. 6:—.

32. Structure investigations of some beryllium materials. By I. Fäldt and G.Lagerberg. 1960. 15 p. Sw. cr. 6:—.

33. An emergency dosimeter for neutrons. By J. Braun and R. Nilsson. 1960.32 p. Sw. cr. 6:—.

34. Theoretical calculation of the effect on lattice parameters of emptyingthe coolant channels in a D2O-moderated and cooled natural uraniumreactor. By P. Weisglas. 1960. 20 p. Sw. cr. 6:—.

35. The multigroup neutron diffusion equations/1 space, dimension. By S.Linde. 1960. 41 p. Sw. cr. 6:—.

36. Geochemical prospecting of a uraniferous bog deposit at Masugnsbyn,Northern Sweden. By G. Armands. 1961. 48 p. Sw. cr. 6:—.

37. Spectrophotometric determination of thorium in low grade minerals andores. By A.-L. Arnfelt and I. Edmundsson. 1960. 14 p. Sw. cr. 6:—.

38. Kinetics of pressurized water reactors with hot or cold moderators. ByO. Norinder. 1960. 24 p. Sw. cr. 6:—.

39. The dependence of the resonance on the Doppler effect. By J. Rosén.1960. 19 p. Sw. cr. «:—.

40. Measurements of the fast fission factor (.<•) in UOj-elemenls. By O. Ny-lund. 1961. Sw. cr. 6:—.

44. Hand monitor for simultaneous measurement of alpha and beta conta-mination. By I. O. Andersson, J. Braun and B. Söderlund. 2nd rev. ed.1961. Sw. cr. 6:—.

45. Measurement of radioactivity in the human body. By I. O . Anderssonand I. Nilsson. 1961. 16 p. Sw. cr. 6:—.

46. The magnetisation of MnB and its variation with temperature. By N.Lundquist and H. P. Myers. 1960. 19 p. Sw. er. 6:—.

47. An experimental study of the scattering of slow neutrons from HjO andD2O. By K. E. Larsson, S. Holmryd and K. Otnes. 1960. 29 p. Sw. cr. 6:—.

48. The resonance integral of thorium metal rods. By E. Hellstrand and J.Weitman. 1961. 32 p. Sw. cr. 6:—.

49. Pressure tube and pressure vessels reactors; certain comparisons. By P.H. Margen, P. E. Ahlström and B. Pershagen. 1961. 42 p. Sw. cr. 6:—.

50. Phase transformations in a uranium-zirconium alloy containing 2 weightper cent zirconium. By G. Lagerberg. 1961. 39 p. Sw. cr. 6:—.

51. Activation analysis of aluminium. By D. Brune. 1961. 8 p. Sw. cr. 6:—.52. Thermo-technical data for D2O. By E. Axblom. 1961. 14 p. Sw .cr. 6:—.53. Neutron damage in steels containing small amounts of boron. By H. P.

Myers. 1961. 23 p. Sw. cr. 6:—.54. A chemical eight group separation method for routine use in gamma

spectrometric analysis. I. Ion exchange experiments. By K. Samsahl.1961. 13 p. Sw. cr. 6:—.

55. The Swedish zero power reactor R0. By Olof Landergård, Kaj Cavallinand Georg Jonsson. 1961. 31 p. Sw. cr. 6:—.

56. A chemical eight group separation method for routine use in gammaspectrometric analysis. I I . Detailed analytical schema. By K. Samsahl.18 p. 1961. Sw. cr. 6:—.

57. Heterogeneous two-group diffusion theory for a finite cylindrical reactor.By Alf Jonsson and Göran Näslund. 1961. 20 p. Sw. cr. 6:—.

58. Q-values for (n, p) and (n, a) reactions. By J. Konijn. 1961. 29 p. Sw. cr.6:—.

59. Studies of the effective total and resonance absorption cross section forzircaloy 2 and zirconium. By E. Hellstrand, G. Lindahl and G. Lundgren.1961.26 p. Sw. cr. 6:—.

60. Determination of elements in normal and leukemic human whole bloodby neutron activation analysis. By D. Brune, B. Frykberg, K. Samsahl andP. O. Wester. 1961. 16 p. Sw. cr. 6:—.

61. Comparative and absolute measurements of 11 inorganic constituents of38 human tooth samples with gamma-ray spectrometry. By K. Samsahland R. Söremark. 19 p. 1961. Sw. cr. 6:—.

62. A Monte Carlo sampling technique for multi-phonon processes. By ThureHögberg. 10 p. 1961. Sw. cr. 6 ™ .

63. Numerical integration of the transport equation for infinite homogeneousmedia. By Rune Håkansson. 1962. 15 p. Sw. cr. 6:—.

64. Modified Sucksmith balances for ferromagnetic and paramagnetic mea-surements. By N. Lundquist and H. P. Myers. 1962. 9 p. Sw. cr. 6:—.

65. Irradiation effects in strain aged pressure vessel steel. By M. Grounesand H. P. Myers. 1962. 8 p. Sw. cr. 6:—.

66. Critical and exponential experiments on 19-rod clusters (R3-fuel) in heavywater. By R. Persson, C-E. Wikdahl and Z. Zadwörski. 1962. 34 p. Sw. cr.6:~~.

67. On the calibration and accuracy of the Guinier camera for the deter-mination of interplanar spacings. By M. Möller. 1962. 21 p. Sw. cr. 6:—.

68. Quantitative determination of pole figures with a texture goniometer bythe reflection method. By M. Möller. 1962. 16 p. Sw. cr. 6:—.

69. An experimental study of pressure gradients for flow of boiling water ina vertical round duct, Part I. By K. M. Becker, G. Hernborg ana M. Bode.1962. 46 p. Sw. cr. 6:—.

70. An experimental study of pressure gradients for flow of boiling water ina vertical round duct. Part I I . By K.M. Becker, G. Hernborg andM. Bode.1962. 32 p. Sw. cr. 6:—.

71.

72.

73.

74.

75.

76.

77.

78.

79.

80.

81.

82.

83.

84.

85.

86.

87.

88.

89.

90.

91.

92.

93.

94.

95.

96.

97.

98.

99.

100

101

102

103.

104.

The space-, lime- and energy-distribution of neutrons from a pulsedplane source. By A. Claesson. 1962. 16 p. Sw. cr. 6:—.One-group perturbation theory applied to substitution measurements withvoid. By R. Persson. 1962. 21 p. Sw. cr. 6:—.Conversion factors. By A. Amberntson and S-E. Larsson. 1962. 15 p. Sw.cr. 10:—.Burnout conditions for flow of boiling water in vertical rod clusters.By Kurt M. Becker. 1962. 44 p. Sw. cr. 6-.—.Two-group current-equivalent parameters for control rod cells. Autocodeprogramme CRCC. By O. Norinder and K. Nyman. 1962. 18 p. Sw. cr.

On the electronic structure of MnB. By N. Lundquist. 1962. 16 p. Sw. cr6:—.The resonance absorption of uranium metal and oxide. By E. Hellstrandand G. Lundgren. 1962. 17 p. Sw. cr. 6:—.Half-life measurements of 'He, " N , »O, *>F, »Al, "Se™ and ™Ag. By J.Konijn and S. Malmskog. 1962. 34 p. Sw. cr. 6:—.Progress report for period ending December 1961. Department for ReactorPhysics. 1962. 53 p. Sw. cr. 6:—.Investigation of the 800 keV peak in the gamma spectrum of SwedishLaplanders. By I. O. Andersson, I. Nilsson and K. Eckerslig. 1962. 8 p.Sw. cr. 6:—.The resonance integral of niobium. By E. Hellstrand and G. Lundgren.1962. 14 p. Sw. cr. 6:—.Some chemical group separations of radioactive trace elements. By K.Samsahl. 1962. 18 p. Sw. cr. 6:—.Void measurement by the (y, n) reactions. By S. Z. Rouhani. 1962. 17 P.Sw. cr. 6:—.Investigation of the pulse height distribution of boron trifluoride pro-portional counters. By I. O. Andersson and S. Malmskog. 1962. 16 p.Sw. cr. 6:—.An experimental study of pressure gradients for flow of boiling waterin vertical round ducts. (Part 3). By K. M. Becker, G. Hernborg and M .Bode. 1962. 29 p. Sw. cr. 6:—.An experimental study of pressure gradients for flow of boiling waterin vertical round ducts. (Part 4). By K. M. Becker, G. Hernborg and M.Bode. 1962. 19 p. Sw. cr. 6:—.

Measurements of burnout conditions for flow of boiling wafer in verticalround ducts. By K. M. Becker. 1962. 38 p. Sw. cr. 6:—.

Cross sections for neutron inelastic scattering and (n, 2n) processes. ByM. Leimdörfer, E. Bock and L. Arkeryd. 1962. 225 p. Sw. cr. 10:—.On the solution of the neutron transport equation. By S. Depfcen. 1962.43 p. Sw. cr. 6:—.Swedish studies on irradiation effects in structural materials. By M.Grounes and H. P. Myers. 1962. 11 p. Sw. cr. 6:—.The energy variation of the sensitivity of a polyethylene moderated BFjproportional counter. By R. Fräki, M. Leimdörfer and S. Malmskog. 1962.12 p. Sw. cr. 6:—.

The backscatteriniM. Leimdörfer. 19<

I of gamma radiation from plane concrete walls. By2. 20 p. Sw. cr. 6:—.

The backscattering of gamma radiation from spherical concrete walls. ByM. Leimdörfer. 1962. 16 p. Sw. cr. 6:—.

Multiple scattering of gamma radiation in a spherical concrete wallroom. By M. Leimdörfer. 1962. 18 p. Sw. cr. 6:—.

The paramagnetism of Mn dissolved in « and ft brasses. By H. P. Myersand R. Weslin. 1962. 13 p. Sw. cr. 6:—. r

Isomorfic substitutions of calcium by strontium in calcium hydroxy-apatite. By H. Christensen. 1962. 9 p. Sw. cr. 6:—.A fast time-to-pulse height converter. By O. Aspelund. 1962. 21 p. Sw. cr.6:—.Neutron streaming in D2O pipes. By J. Braun and K. Randen. 1962.41 p. Sw. cr. 6:—.The effective resonance integral of thorium oxide rods. By J. Weitman.1962. 41 p. Sw. cr. 6:—.

1. Measurements of burnout conditions for flow of boiling water in verticalannuli. By K. M. Becker and G. Hernborg. 1962. 41 p. Sw. cr. 6:—.

. Solid angle computations for a circular radiator and a circular detector.By J. Konijn and B. Tollander. 1963. 6 p. Sw. cr. 8:—.

:. A selective neutron detector in the keV region utilizing the "F(n, y)!0Freaction. By J. Konijn. 1963. 21 p. Sw. cr. 8:—.

Anion-exchange studies of radioactive trace elements in sulphuric acidsolutions. By K. Samsahl. 1963. Sw. cr. 8:—.Problems in pressure vessel. Design and manufacture. By O. Hellströmand Ragnar Nilson. 1963. Sw. cr. 8:—.

Förteckning över publicerade AES-rapporter

1. Analys medelst gamma-spektrometri. Av Dag Brune. 1961. 10 s. Kr 6:—.

2. Besträlningsförändringar och neutronatmosfär i reaktortrycklankar —några synpunkter. Av M. Grounes. 1962. 33 s. Kr 6:—.

Additional copies available at the library of AB Atomenergi, Studsvik, Nykö-ping, Sweden. Transport microcards of the reports are obtainable throughthe International Documentation Center, Tumba, Sweden.

EOS-tryckerierna, Stockholm 1963